Energy From Thorium Discussion Forum

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PostPosted: Feb 14, 2010 3:33 pm 
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Cyril R wrote:
So, with an efficient bi modal reactor, do you think there is not much of a case for a few synergetic fast reactors? If you start up on near natural U-235/U238 concentration, then there's not much point in breeding U-233 with fast reactors, whether chloride or fluoride. What about transuranics burning from existing spent nuclear fuel?

Also, with thorium in the fuel salt, the reprocessing tends to be more expensive and complicated - will this be a problem if you're going to rely heavily on continuous reprocessing?

Thanks Cyril -- I think you pretty much answered your own questions...

The last one may be the deciding factor: Some folks prefer avoiding on-line fuel processing, which definitely fovours reactor types that are much less sensitive to FP contamination -- namely fast reactors.

I would only reiterate that the fuel volumes we are talking about are small, as is the processing equipment.
So I'm not really sure why it should be all that expensive.

And even with fast reactors, we expect the fuel to be processed at some point anyway -- so reprocessing equipment will be required, one way or another -- sooner or later.
Is there likely to be a big cost difference between these processing options ? ...I don't think so (but that's just my opinion).

Fuel volume is especially well minimised with well-thermalised reactor designs, and with high operating temperatures (to get the same power output at similar fuel flow rates through the primary HX).


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PostPosted: Feb 14, 2010 5:07 pm 
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I'm not sure if I get it yet. It seems that in your design, when running a small salt volume in single fluid, not much thorium can be put in the fuel salt, because there's too much Pa capture. So Th breeding is tiny and you simply rely on the fast portion of the neutrons breeding from 238-U, effectively running on Pu. With more Th in the fuel salt, Pa extraction will likely be necessary, causing political issues of proliferation. Might as well go for two fluid then (put the thorium in the calandria in solution or suspension). Or, don't bother with thorium at all in the first designs, and run as close to breakeven as possible on LEU?


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PostPosted: Feb 14, 2010 6:01 pm 
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Cyril R wrote:
I'm not sure if I get it yet. It seems that in your design, when running a small salt volume in single fluid, not much thorium can be put in the fuel salt, because there's too much Pa capture. So Th breeding is tiny and you simply rely on the fast portion of the neutrons breeding from 238-U, effectively running on Pu. With more Th in the fuel salt, Pa extraction will likely be necessary, causing political issues of proliferation. Might as well go for two fluid then (put the thorium in the calandria in solution or suspension). Or, don't bother with thorium at all in the first designs, and run as close to breakeven as possible on LEU?

I think that's an excellent assessment.

As you say, Pa extraction is a great big political proliferation issue, which we would rather avoid.

One way to deal with this might be some kind of out-of-core buffer, that doesn't do anything other than provide decay time for Pa233, prior to dumping back into the reactor.
This is the approach proposed for the PB-AHTR, where Th pebbles would be buffered outside the reactor.
Such a buffer would leave the new U233 mixed with U238 at all times.
One would have to look at optimising the size of the buffer, such that it wouldn't affect the scale of the fuel purification equipment too much.
The latter is really only concerned with just the fuel circulating through the reactor -- so it shouldn't be affected very much by a bunch of stuff sitting idle in a buffer.
What do you think ?


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PostPosted: Feb 14, 2010 6:51 pm 
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jaro wrote:
Cyril R wrote:
I'm not sure if I get it yet. It seems that in your design, when running a small salt volume in single fluid, not much thorium can be put in the fuel salt, because there's too much Pa capture. So Th breeding is tiny and you simply rely on the fast portion of the neutrons breeding from 238-U, effectively running on Pu. With more Th in the fuel salt, Pa extraction will likely be necessary, causing political issues of proliferation. Might as well go for two fluid then (put the thorium in the calandria in solution or suspension). Or, don't bother with thorium at all in the first designs, and run as close to breakeven as possible on LEU?

I think that's an excellent assessment.

As you say, Pa extraction is a great big political proliferation issue, which we would rather avoid.

One way to deal with this might be some kind of out-of-core buffer, that doesn't do anything other than provide decay time for Pa233, prior to dumping back into the reactor.
This is the approach proposed for the PB-AHTR, where Th pebbles would be buffered outside the reactor.
Such a buffer would leave the new U233 mixed with U238 at all times.
One would have to look at optimising the size of the buffer, such that it wouldn't affect the scale of the fuel purification equipment too much.
The latter is really only concerned with just the fuel circulating through the reactor -- so it shouldn't be affected very much by a bunch of stuff sitting idle in a buffer.
What do you think ?


Sounds good. One thing that's important is how long you can circulate the thorium before sending to the buffer. The Pa-233 has a half life of almost a month so it would have to be stored quite some time before reinsertion. If circulation time (or rate, it there's a continuous process) can be long (slow rate) then it could work well at low cost. If it is short (fast rate), seems like a lot has to be buffered, that means a lot more uranium fluoride inventory. Is that a problem?


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PostPosted: Feb 14, 2010 7:00 pm 
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Cyril R wrote:
If circulation time (or rate, it there's a continuous process) can be long (slow rate) then it could work well at low cost. If it is short (fast rate), seems like a lot has to be buffered, that means a lot more uranium fluoride inventory. Is that a problem?

It would be a problem if we were using enriched fuel -- the more its enriched, the bigger the problem.

With NU & Th, not a problem.


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PostPosted: Feb 14, 2010 10:23 pm 
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jaro wrote:
As you say, Pa extraction is a great big political proliferation issue, which we would rather avoid.


Per Peterson doesn't think so, and he's an expert on the subject. I wouldn't be so hasty in my judgement about the risk associated with Pa extraction.


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PostPosted: Feb 15, 2010 2:07 am 
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In a Liquid Chloride(Cl37) medium, the neutron energy and neutron production will be so high that it can burn all the actinides. Removing Pa233 for avoiding a small wastage is not worthwhile. There is no graphite problem either in fast reactors. Corrosion problem will be so much less than that of fluoride salts that materials of present fast reactors are likely to work satisfactorily. The present generation of fast reactors need a stable salt coolant and the Liquid Chloride reactors provide in addition a fluid fuel too, which helps remove volatile fission products for improved neutron efficiency.
I wish people with spare fissile isotopes, reactor grade plutonium as well as weapon materials would work on fast reactors (including Liquid Chloride fueled) to make use of it for energy purposes. Only Russia out of these seems to have fast reactors on their mind at present. India and China among others are also making efforts. US, UK and France should follow their lead.


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PostPosted: Feb 15, 2010 3:01 am 
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Kirk Sorensen wrote:
jaro wrote:
As you say, Pa extraction is a great big political proliferation issue, which we would rather avoid.


Per Peterson doesn't think so, and he's an expert on the subject. I wouldn't be so hasty in my judgement about the risk associated with Pa extraction.



Per Peterson: "But LFTRs will still require effective IAEA safeguards to assure that the state cannot divert material without detection, and will have similar non-proliferation issues with other fission reactors"


The impact of such IAEA oversight is minimize in the breeder/burner model where of few breeders provide U233/U232 to many burners. Pers system follows the breeder/burner model because his TRISO fuel fabrication is done by centralize trusted facility.

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PostPosted: Feb 15, 2010 4:14 am 
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I'm not worried about the risk of proliferation of Pa extraction, but about what regulators are going to think and do about it. If Per and Kirk are correct then all the better, we can do with a small Pa storage. Having a much larger Pa storage (diluted with U-238) sounds acceptable though.


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PostPosted: Feb 15, 2010 1:06 pm 
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By this do you mean to extract all the Pa as it gets generated and then let it decay in 238U so that you have 13% 233U? If so, then you basically have a plutonium reactor not a 233U reactor since you are adding 6 238U atoms for each 233U. You will need to eventually be burning 6 239Pu atoms for every 233U atom.


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PostPosted: Feb 15, 2010 1:40 pm 
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Cyril R wrote:
I'm not worried about the risk of proliferation of Pa extraction, but about what regulators are going to think and do about it. If Per and Kirk are correct then all the better, we can do with a small Pa storage. Having a much larger Pa storage (diluted with U-238) sounds acceptable though.



The devil is in the details.

Let’s talk details


Anytime the core salt uranium fissile isotopic composition is actively modified by human intervention regardless if it involves Pa storage fissile retrieval and /or on-site reprocessing, IAEA on-site inspectors will independently verify that the appropriate IAEA denaturing specifications of the core salt as a whole is compliant with IAEA rules.

In the case where the core salt design ops for 88% U238 denaturing of U233, this principle is applied so that regardless of the nuclear waste mix present in the core salt (U234,u236,np237, etc…) 88% u238 is always present as a denaturant of U233.

Even if a fully automated core salt reprocessing technology is implemented, on-site IAEA inspection will be required and an appropriate human inspection interface will be provided.

The self protection characteristic of the core salt does not bear upon the IAEA denaturing specification and is an independent issue.

This inspection requirement may preclude unattended underground deployment to facilitate this on-site IAEA inspection regime.

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PostPosted: Feb 15, 2010 1:59 pm 
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Lars wrote:
By this do you mean to extract all the Pa as it gets generated and then let it decay in 238U so that you have 13% 233U? If so, then you basically have a plutonium reactor not a 233U reactor since you are adding 6 238U atoms for each 233U. You will need to eventually be burning 6 239Pu atoms for every 233U atom.

The important thing is to gain some molten salt reactor experience, in anticipation of a more permissive regulatory regime sometime in the future (we hope), when fissile dilution ("denaturing") will no longer be required.

Putting it another way: Should we abandon all MSR development just because we can't have a "pure thorium reactor" (with un-denatured U233) under current non-proliferation policy ?

If the answer is "no", then perhaps it might make sense to look at MSR variants that can operate successfully under current restrictions.


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PostPosted: Feb 15, 2010 2:08 pm 
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Quote:
Putting it another way: Should we abandon all MSR development just because we can't have a "pure thorium reactor" (with un-denatured U233) under current non-proliferation policy ?

If the answer is "no", then perhaps it might make sense to look at MSR variants that can operate successfully under current restrictions.


The IAEA provides two types of U233 denaturing as follows:

88% U238 to 12% U233

and

1% U232 to 99% U233

The pure U233 fuel cycle can be implemented with using 1% U232 as the denaturant.

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PostPosted: Feb 15, 2010 2:15 pm 
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Fair enough but I at least do not fully understand the current restrictions as regards 233U inside the core. My guess is that the rules for this have not been developed yet. There is the denatured MSR design that the British declared to be the most proliferation resistant reactor design. If we do have to run denatured then the advantages of a heavy water reactor become much stronger.

I interpreted some of Dr. Peterson's comments as meaning that we would likely be allowed to have relatively rich 233U in the core as long as it was protected by plenty of radiation. This makes some sense to me. The difference in the current treatment of spent fuel versus isolated plutonium for MOX also suggests that we are given anti-proliferation credit for keeping fissile together with plenty of radiation.


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PostPosted: Feb 15, 2010 2:37 pm 
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Quote:
I interpreted some of Dr. Peterson's comments as meaning that we would likely be allowed to have relatively rich 233U in the core as long as it was protected by plenty of radiation. This makes some sense to me. The difference in the current treatment of spent fuel versus isolated plutonium for MOX also suggests that we are given anti-proliferation credit for keeping fissile together with plenty of radiation.


Fuel self protection applies as long as there is no opportunity or possibility of radiation reduction through reprocessing. If you load the core salt and LEAVE IT ALONE, then self protection through radiation is applicable.

In the Peterson system the TRISO fuel format excludes human intervention until the fuel reaches the reprocessing center. There is no on-site reprocessing in the Peterson system. Because of this fuel protection by the TRISO format, because humans cannot modify the elemental components of the TRISO seed and blanket pebbles, Peterson can run a pure U233 fuel cycle.


By the way, 1% U232 denaturing does not depend of self protection from radiation as its anti-proliferation mechanism. U232 denaturing uses the same physical principles and natural laws that make reactor grade plutonium proliferation safe. Reprocessing may be possible if U232 concentrations are kept above 1%.

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